Important requirements for materials used in boiling water nuclear reactor construction include low absorption for thermal neutrons, corrosion and stress-corrosion resistance and mechanical strength. Zirconium-base alloys sufficiently satisfy these requirements that they are widely used for such purposes, "Zircaloy-2" (containing about 1.5 percent tin, 0.15 percent iron, 0.1 percent chromium, 0.05 percent nickel and 0.1 percent oxygen) and "Zircaloy-4" (containing substantially no nickel but otherwise similar to Zircaloy-2) being two of the important commercial alloys commonly finding such use. These alloys, however, are not nearly all that one would desire, particularly in respect to useful service life, despite many efforts of others during the past two decades to improve them. Mainly, these efforts have been aimed at improving corrosion resistance and usually this has involved changes in composition. Thus, in U.S. Pat. No. 3,005,706, it is proposed that from 0.03 to 1.0 percent of beryllium be added to zirconium alloys intended for use in conventional boilers, boiling water reactors and similar apparatus. Similarly, in U.S. Pat. Nos. 3,261,682 and 3,150,972, cerium and/or yttrium, and calcium, respectively, are proposed as zirconium alloy additions in like proportions for the same purpose. Accounts and reports of the results of such compositional changes are sparse, however, and the present commercial alloys do not include any of these additional constituents.
The literature in this field, however, contains little concerning efforts to improve upon the mechanical strength of zirconium-base alloys and particularly the load-carrying capacity of fuel cladding and other reactor parts subjected to prolonged exposure to typical boiling water reactor conditions. This is in spite of the fact that it has long been general knowledge that slow strain rate ductility of these alloys is lost to a great extent as a result of radiation exposure over periods of a year or more. The problem of premature termination of service life because of fast neutron radiation-induced embrittlement is particularly aggravated in the case of nuclear fuel containment channels and tubes or cladding. The natural swelling of the fuel as it is burned produces high localized stresses leading to stress-corrosion cracking of the cladding at a time before corrosion of the type described in the above patents might normally necessitate cladding replacement.